Planning and supervision of reactor defueling using discrete event techniques

New fuel handling and conditioning activities for the defueling of the Experimental Breeder Reactor II are being performed at Argonne National Laboratory. Research is being conducted to investigate the use of discrete event simulation, analysis, and optimization techniques to plan, supervise, and perform these activities in such a way that productivity can be improved. The central idea is to characterize this defueling operation as a collection of interconnected serving cells, and then apply operational research techniques to identify appropriate planning schedules for given scenarios. In addition, a supervisory system is being developed to provide personnel with on-line information on the progress of fueling tasks and to suggest courses of action to accommodate changing operational conditions. This paper provides an introduction to the research in progress at ANL. In particular, it briefly describes the fuel handling configuration for reactor defueling at ANL, presenting the flow of material from the reactor grid to the interim storage location, and the expected contributions of this work. As an example of the studies being conducted for planning and supervision of fuel handling activities at ANL, an application of discrete event simulation techniques to evaluate different fuel cask transfer strategies is given at the end of the paper

RFQ-based, transportable, high-resolution, neutron radiography system concept

A preliminary design for a high-resolution transportable neutron radiography system concept has been developed. The system requirement has been taken to be a thermal neutron flux of 10{sup 6} N/(cm{sup 2}- sec) with an L/D of 100. The approach is to use an accelerator-driven neutron source, with a radiofrequency quadrupole (RFQ) as the primary accelerator component. Initial concepts for all of the major components of the system have been developed, and selected key parts have been examined further. An overview of the system design is presented, together with brief summaries of the concepts for the ion source, LEBT, RFQ, HEBT, target, moderator, collimator, image collection, power, cooling, vacuum, structure, robotics, control system, data analysis, transport vehicle, and site support. More detailed studies completed for the RFQ and moderator designs, and issues identified during the course of the work, are described

Neutronics calculations for the TFTR neutron calorimeter

Neutronics anayses have been performed for an adiabatic neutron calorimeter consisting of a pure hydrocarbon moderator located just outside of the vacuum vessel of the Tokamak Fusion Test Reactor. One and two-dimensional neutronic analyses show that the incident fusion neutron fluence can be determined to +- 10% uncertainty by simply integrating measured temperature profiles along the central radial axis (and assuming negligible error in the temperature measurement). The +- 10% uncertainty is found to be due to gamma rays produced by inelastic scattering and exothermic capture reactions in the moderator and vacuum vessel. The perturbing effects due to the toroidal field coils and due to gamma rays entering the sides of the calorimeter are shown to be negligible in the region of the central axis

Time series analysis of nuclear instrumentation in EBR-II

Results of a time series analysis of the scaler count data from the 3 wide range nuclear detectors in the Experimental Breeder Reactor-II are presented. One of the channels was replaced, and it was desired to determine if there was any statistically significant change (ie, improvement) in the channel`s response after the replacement. Data were collected from all 3 channels for 16-day periods before and after detector replacement. Time series analysis and statistical tests showed that there was no significant change after the detector replacement. Also, there were no statistically significant differences among the 3 channels, either before or after the replacement. Finally, it was determined that errors in the reactivity change inferred from subcritical count monitoring during fuel handling would be on the other of 20-30 cents for single count intervals

Source-to-incident-flux relation in a tokamak blanket module

The next-generation tokamak experiments, including the Tokamak Fusion Test Reactor (TFTR), will utilize small blanket modules to measure performance parameters such as tritium breeding profiles, power deposition profiles, and neutron flux profiles. Specifically, a neutron calorimeter (simply a neutron moderating blanket module) which allows one to infer the incident 14 MeV flux based on measured temperature profiles has been proposed for TFTR. The problem addressed here is how to relate this total scalar flux to the fusion neutron source; this relation is necessary since the calorimeter is proposed as a total fusion energy monitor. The methods and assumptions presented here will be valid for the TFTR Lithium Breeding Module (LBM), as well as other modules on larger tokamak reactors

Beam characterization at the Neutron Radiography Facility (NRAD)

An ongoing project to characterize the neutron beams at the Neutron Radiography Reactor (NRAD) is described in this paper. The effort has consisted of computer modelling with three dimensional diffusion theory to obtain a trail spectrum, foil activation measurements, and the use of SAND-II unfolding code. It was expected and found that diffusion theory will underpredict the fast flux. However, it is claimed that precise characterization of the entire spectrum is not necessary for comparisons among neutron radiography facilities; rather, the use of simple fast neutron indicators should be adequate. A specific example used at NRAD is the U-235/U-238 fission reaction rate ratio. A ratio such as this could be used in the same manner as the classic gold cadmium ratio for interfacility comparisons with regard to fast neutrons. 5 refs

A PC based computerized maintenance system

The present regulatory climate in the research reactor community has made an easily manageable and auditable maintenance system a necessity. We at NRAD have developed a computer-based system that is easy to implement and use, meets all our regulatory and reporting requirements, and is extremely useful to us in our daily operations. The system, developed at the NRAD reactor facility at Argonne National Laboratory in Idaho Falls, Idaho, uses DBASE-III coupled with C language routines, written for specific purposes. It is a menu-driven system that can be mastered in a short period of time and maintained with only a few hours of computer operation per month. It uses three computer processes: job scheduling, file updating, and report preparation, to produce schedules, work orders, and miscellaneous report forms. The heart of the system is an IBM PC with a 10 MB hard disk, providing adequate data storage capacity for a facility the size of NRAD. The computer is totally dedicated to the maintenance system, thus guarding against inadvertent loss of, or damage to, data files. Computer operator training time is minimized by the menu driven program. Multiple operators can share the computer operation responsibilities, and maintain the system with only 12 to 16 hours of computer operation per month. The system is adaptable to almost any facility, and can be altered and expanded to satisfy changing requirements. 7 figs

The Neutron Radiography Reactor (NRAD)

The Neutron Radiography Reactor (NRAD) operated by Argonne National Laboratory is described in this paper. NRAD was designed to allow radiography of highly absorbing reactor fuel assemblies in the vertical position on the routine basis. 7 figs

Moderator/collimator for a proton/deuteron linac to produce a high-intensity, high-quality thermal neutron beam for neutron radiography

Reactor based high resolution neutron radiography facilities are able to deliver a well-collimated (L/D {ge}100) thermal flux of 10{sup 6} n/cm{sup 2}{center_dot}sec to an image plane. This is well in excess of that achievable with the present accelerator based systems such as sealed tube D-T sources, Van der Graaff`s, small cyclotrons, or low duty factor linacs. However, continuous wave linacs can accelerate tens of milliamperes of protons to 2.5 to 4 MeV. The MCNP code has been used to analyze target/moderator configurations that could be used with Argonne`s Continuous Wave Linac (ACWL). These analyses have shown that ACWL could be modified to generate a neutron beam that has a high intensity and is of high quality

Neutron flux enhancement in the NRAD reactor

In 1987 a series of experiments were conducted at the NRAD reactor facility at Argonne National Laboratory - West (ANL-W) to investigate the possibility of increasing the thermal neutron content at the end of the reactor's east beam tube through the use of hydrogenous flux traps. It was desired to increase the thermal flux for a series of experiments to be performed in the east radiography cell, in which the enhanced flux was required in a relatively small volume. Hence, it was feasible to attempt to focus the cross section of the beam to a smaller area. Two flux traps were constructed from unborated polypropylene and tested to determine their effectiveness. Both traps were open to the entire cross-sectional area of the neutron beam (as it emerges from the wall and enters the beam room). The sides then converged such that at the end of the trap the beam would be 'focused' to a greater intensity. The differences in the two flux traps were primarily in length, and hence angle to the beam as the inlet and outlet cross-sectional areas were held constant. It should be noted that merely placing a slab of polypropylene in the beam will not yield significant multiplication as neutrons are primarily scattered away